Regulatory Guide 1.33: Difference between revisions

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{{Adams
{{Adams
| number = ML13109A458
| number = ML003739995
| issue date = 06/13/2013
| issue date = 02/28/1978
| title = Rev 3, Quality Assurance Program Requirements (Operation).
| title = Quality Assurance Program Requirements (Operation)
| author name =  
| author name =  
| author affiliation = NRC/RES
| author affiliation = NRC/RES
Line 9: Line 9:
| docket =  
| docket =  
| license number =  
| license number =  
| contact person = Rodriguez-Luccioni H L
| contact person =  
| case reference number = DG-1300
| document report number = RG-1.33, Rev 2
| document report number = RG-1.033, Rev 3
| package number = ML13109A437
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 6
| page count = 8
}}
}}
{{#Wiki_filter:Written suggestions regarding this guide or development of new guides may be submitted through the NRC's public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.    Electronic copies of this regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRC's public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/.  The regulatory guide is also available through the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession No. ML13109A458.  The regulatory analysis may be found in ADAMS under Accession No. ML13109A459 and the staff responses to the public comments on DG-1300 may be found under ADAMS Accession No. ML13109A467.
{{#Wiki_filter:Revlklon 2 U.S. NUCLEAR REGULATORY COMMISSION                                                                                February 1978 REGULATORY GUIDE
                                      OFFICE OF STANDARDS DEVELOPMENT
                                                                      REGULATORY GUIDE 1.33 QUALITY ASSURANCE PROGRAM REQUIREMENTS
                                                                                  (OPERATION)


U.S. NUCLEAR REGULATORY COMMISSION June 2013 Revision 3 REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH
==A. INTRODUCTION==
REGULATORY GUIDE 1.33 (Draft was issued as DG-1300, dated January 2013)  
ANS-3.2, "Administrative Controls and Quality As surance for the Operational Phase of Nuclear Power Appendix B, "Quality Assurance Criteria for Nu                                        Plants," by the American National Standards Insti clear Power Plants and Fuel Reprocessing Plants," to                                      tute on February 19, 1976.
QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION)
 
10 CFR Part 50, "Licensing of Production and Utili zation Facilities," establishes quality assurance re                                          There had been some uncertainty with regard to the quirements for the operation of nuclear power plant                                        NRC staff's position when a regulatory guide en safety-related structures, systems, and components.                                        dorses, as an acceptable method, the "guidelines" as This regulatory guide describes a method acceptable                                        well as the "requirements" included in a standard.
 
to the NRC staff for complying with the Commis                                            The NRC staff has evaluated the guidelines contained sion's regulations with regard to overall quality as                                        in N18.7-1976/ANS-3.2 with respect to importance surance program requirements for the operation phase                                      to safety. Revision I of this regulatory guide clarified of nuclear power plants. The Advisory Committee on                                          the NRC staff's position on the "requirements" and Reactor Safeguards has been consulted concerning                                            "guidelines" included in ANSI N18.7-1976/ANS
this guide and has concurred in the regulatory                                              3.2. Where conformance to the recommendations of position.                                                                                  this regulatory guide is indicated in an application without further qualification, this indicates the appli
 
==B. DISCUSSION==
cant will comply with the "requirements" of ANSI
                                                                                            N18.7-1976/ANS-3.2, as supplemented or modified Subcommittee ANS-3,' Reactor Operations, of the                                        by the regulatory position of this guide.
 
American Nuclear Society Standards Committee de veloped ANSI N18.7-1972, which contained criteria for                                        Section 1, "Scope," of ANSI NI8.7-1976/ANS
administrative controls for nuclear power plants dur                                      3.2 states that this standard contains criteria for ad ing operation. This standard, along with ANSI                                            ministrative controls and quality assurance for nu N45.2-1971, "Quality Assurance Program Require                                            clear power plants during the operational phase of ments for Nuclear Power Plants," was endorsed by                                          plant life and that this phase is generally considered Regulatory Guide 1.33. The dual endorsement was                                           to commence with initial fuel loading, except for cer necessary in order for the guidance contained in the                                      tain preoperational activities. In this regard, a sepa regulatory guide to be consistent with the require                                        rate regulatory guide addressing the quality assurance ments of Appendix B to 10 CFR Part 50; however,                                          program for the preoperational phase will be issued.
 
this dual endorsement caused some confusion among                                        Other regulatory guides may be issued or this regula users. To clarify this situation, ANSI N18.7-1972                                          tory guide may be revised, if necessary, to amplify was revised so that a single standard would define the                                  the general requirements contained in this standard.
 
general quality assurance program "requirements"
for the operation phase. This revised standard was                                            Appendix A to this guide has been further revised approved by the American National Standards Com                                          as a result of additional' comments received on the mittee NI8, Nuclear Design Criteria. It was sub                                          guide and additional staff review.
 
sequently approved and designated N18.7-1976/
                                                                                          ,Copies may be obtained from American Nuclear Society, 555
-*Lines indicate substantive changes from previous issue.                                North Kensington Avenue, La Grange Park, Illinois 60525.
 
USNRC REGULATORY GUIDES                                        Comments shoukl      be sent to the Secretary of the Commission. US. Nuclear Regu latory Commission, Washington, D.C.        20555, Attention: Docketing and Service Regulatory Guides ore issued to describe and make available to the Public methods acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evaluating specific problems The guides are issued in the following ten broad divisions or postulated accidents, or to Provide guidance to applicants. Regulatory Guides awt not substitutes for regulations, and compliance with them is not required.          1.  Power Reaclors                          6. TProducts Methods and solutions different from those set out in the guides will be accept.      2.  Research and Test Reactors              7. Trancsortation aWle if they provide a basis for the findings requisite to the issuance or continuance 3.  Fuels and Materials Facilities          a. Occupational Health o fa p ecr mitor license aby th eCo mni, son.                                          4.  Environm etntl and Siting               
 
===9. Antitrust Review===
                                                                                        5.  Materials and Plant Protection          t0. General Comments end suggestions for improvements in these guides we encouraged at all        Requests for single copies of issued guides iwhich may be reiroduced) or for place tires, and guides will be revised, as appropriate, to accomnmodate comments and        ment on an automatic distribution list for single copies of future guides in specific
"toreflect new information or experience. This guide was revised as a result of        divisions should be made in writing to the US. Nuclear Regulatory Commission.


==A. INTRODUCTION==
eubltantive comments received from the Public and additional staff review.              Washington, D.C.      20555. Attention    Director. Division of Document Coistrol.
Purpose  This regulatory guide (RG) describes methods that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for managerial and administrative Quality Assurance (QA) controls to be used for nuclear power plants during operations.
 
==C. REGULATORY POSITION==
and will be endorsed by a regulatory guide The overall quality assurance program require                      upon its approval as an ANSI standard.


Applicable Rules and Regulations This guide describes methods that the NRC staff considers acceptable for complying with the provisions of regulations in 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," of the Code of Federal Regulations (Ref. 1), §50.34(b)(6)(ii), Contents of applications; technical information  and  10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants," (10 CFR Part 52) (Ref. 2) §52.79(a)(27), Contents of applications; technical information in final safety analysis report.  Both sections require compliance with 10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," which, in part, requires the establishment of QA controls for the implementation of managerial and administrative controls to assure safe operation.
ments for the operation phase that are included in ANSI N18.7-1976/ANS-3.2 are acceptable to the                       3. Section 4.3.4, "Subjects Requiring Independ NRC staff and provide an adequate basis for comply              ent Review," Item (3) states, in part, that changes to ing with the quality assurance program requirements            the technical specifications or license amendments re of Appendix B to 10 CFR Part 50, subject to the fol              lated to nuclear safety are required to be reviewed by lowing:                                                          the independent review body prior to implementation.


Related Guidance Guidance for the establishment and execution of QA programs for nuclear power plants during their design and construction is in RG 1.28, "Quality Assurance Program Requirements (Design and Construction)," (Ref. 3).  
It should be noted that proposed changes to technical I. ANSI N18.7-1976/ANS-3.2 requires the prep                specifications or license amendments should be re aration of many procedures to carry out an effective            viewed by the independent review body prior to their quality assurance program. Appendix A, "Typical                submittal to the Commission for approval.


Rev. 3 of RG 1.33, Page 2 Purpose of Regulatory Guides The NRC issues regulatory guides to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agency's regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis fo r the findings required for the issuance or continuance of a permit or license by the Commission.
Procedures for Pressurized Water Reactors and Boil ing Water Reactors," to this regulatory guide should                4. Section 4.5, "Audit Program," of ANSI
be used as guidance to ensure minimum procedural                N18.7-1976/ANS_3.2 states that audits of selected coverage for plant operating activities, including re          aspects of operational phase activities shall be per lated maintenance activities. Appendix A lists typical          formed with a frequency commensurate with their safety-related activities that should be covered by            safety significance and in such a manner as to ensure written procedures but does not provide a complete              that an audit of all safety-related functions is com listing of needed procedures. Many other activities            pleted within a period of 2 years. In amplification of carried out during the operation phase of a nuclear            this requirement, the following program elements power plant require written procedures not included            should be audited at the indicated frequencies:
in Appendix A. Appendix A may also contain proce dures that are not applicable to an applicant because                  a. The results of actions taken to correct de of the configuration of the nuclear power plant. The            ficiencies that affect nuclear safety and occur in facil procedures listed in Appendix A may be combined,                ity equipment, structures, systems, or method of separated, or deleted to conform to the applicant's            operation-at least once per 6 months.


Paperwork Reduction Act This regulatory guide contains information collection requirements covered by 10 CFR Part 50 and 10 CFR Part 52 that the Office of Management and Budget (OMB) approved under OMB control numbers 3150-0011 and 3150-0151, respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an informa tion collection request or requirement unless the requesting document displays a currently valid OMB control number.
procedures plan.                                                      b. The conformance of facility operation to pro
    2. Throughout ANSI NI8.7-1976/ANS-3.2, other                visions contained within the technical specifications documents required to be included as a part of this            and applicable license conditions-at least once per standard are identified at the point of reference. The          12 months.


==B. DISCUSSION==
specific acceptability of these standards listed in ANSI N18.7-1976/ANS-3.2 has been addressed in                          c. The performance, training, and qualifications the latest revision of the following regulatory guides:        of the facility staff-at least once per 12 months.
Reason for Revision This revision (Revision 3) of RG 1.33 endorses ANSI/ANS 3.2-2012, "Managerial, Administrative, and Quality Assurance Controls for Operational Phase of Nuclear Power Plants,"  (Ref. 4).  Revision 2 of RG 1.33 endorsed a previous version of the standard, which was ANS 3.2/ANSI N18.7-1976, "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," (Ref. 5).  The updated standard incorporates operational experience since the original standard was developed, and is better focused on QA of plant operations because information on QA of design and construction was moved to another standard.


Background Revision 2 of RG 1.33 (Ref. 6) endorsed ANS 3.2/ANSI 18.7-1976, but required numerous clarifications or modifications of the standard in the RG's Regulatory Position section. Since then, licensees have obtained NRC approval to use various alternate positions to Revision 2 of RG 1.33.
5. The ANSI Standard          Regulatory Guide                "should") guidelines        (indicated by the verb of ANSI N18.7-1976/ANS-3.2 contained N45.2                    1.28                        in the following sections have sufficient safety impor N45.2.1                  1.37                        tance to be treated the same as the requirements (in N45.2.2                  1.38                        dicated by the verb "shall") of the standard:
        N45.2.3                   1.39 N45.2.4                                                      a. Section 4.4-The guidelines concerning re
                                  1.30
        N45.2.5                                                view activities of the onsite operating organization,
                                  1.94                        except the guideline that refers to screening subjects N45.2.6                    1.58 N45.2.8                                                of potential concern.


In addition, the American Society of Mechanical Engineers (ASME) issued NQA-1, "Quality Assurance Program Requirements for Nuclear Power Plants," (Ref. 7), which was focused on design and construction issues. The NRC has endorsed NQA-1 in 10 CFR 50.55a, "Codes and Standards". The NRC revised a related RG (RG 1.28) to endorse NQA-1-2008 and the NQA-1a-2009 Addenda, "Quality Assurance Requirements for Nuclear Facility Applications."
1.116 N45.2.9                    1.88                              b. Section 5.2.3-The guideline concerning re N45.2. 10                  1.74                        view and updating of standing orders.


ANSI/ANS 3.2-2012 revised ANS 3.2/ANSI 18.7-1976 to remove information related to design and construction to be consistent with NQA-1, and to incorporate the alternate positions approved by the NRC since ANS 3.2/ANSI 18.7-1976 was issued. Revision 3 of RG 1.33 clarifies the distinction of the quality assurance program during design and construction from those managerial and administrative controls implemented during the operational phase of nuclear power plants.
N45.2. I I                1.64 N45.2.13                  1.123                              c. Section 5.2.4-The guideline concerning re N18. 1                                               view, updating, and cancellation of special orders.


Rev. 3 of RG 1.33, Page 3 Harmonization with International Standards The International Atomic Energy Agency (IAEA) has established a series of safety guides and standards constituting a high level of safety for protecting people and the environment. IAEA safety guides present international good practices and increasingly reflect best practices to help users striving to achieve high levels of safety. Pertinent to this regulatory guide, the IAEA Safety Standards, and their Safety Requirement GS-R-3, "The Management System for Facilities and Activities," (Ref. 8), issued in 2006, address administrative and quality assurance controls for the operational phase of nuclear power plants.  This regulatory guide incorporates similar administrative and quality assurance controls for the operational phase and is consistent with the basic safety principles provided in the IAEA Safety Standard.
1.8 N 18.17                    1.17                              d. Section 5.2.7. 1-The guidelines that address N 101.4                    1.54                      adequate design and testing of replacement parts.


Documents Discussed in Staff Regulatory Guidance This regulatory guide endorses the use of one or more voluntary consensus codes or standards developed by external organizations. These codes or standards may contain references to other codes or standards. These references should be considered individually. If a referenced standard has been incorporated separately into NRC regulations, licensees and applicants must comply with that standard as set forth in the regulation. If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the NRC staff for meeting a regulatory requirement as described in the specific regulatory guide.  If a refere nced standard has been neither incorporated into NRC regulations nor endorsed in a regulatory guide, licensees and applicants may consider and use the information in the referenced standard, if appropriate ly justified and consistent with current regulatory practice.
Note: N45.2.12 is discussed in NRC                          e. Section 5.2.13.4-The guideline concerning documents WASH-1283, "Guidance on                        special handling tools and equipment.


C. STAFF REGULATORY GUIDANCE  The requirements included in ANSI/ANS 3.2-2012, "Managerial, Administrative and Quality Assurance Controls for the Operational Phase of Nuclear Power Plants", for implementation during the operation phase of nuclear power plants, are acceptable to the NRC staff and provide an adequate basis for complying with the requirements of Appendix B to 10 CFR Part 50, subject to the following condition on the use of ANSI/ANS 3.2-2012:
Quality Assurance Requirements During Design and Procurement Phase of Nuclear                        f. Section 5.2.19(2)-The guideline for check Power Plants," (Grey Book) and WASH                    ing plant operating procedures during the testing pro
    1309, "Guidance on Quality Assurance Re                gram.


* ANSI/ANS 3.2-2012 requires the preparation of many procedures to carry out an effective QA program. Appendix A of ANSI/ANS 3.2-2012, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," should be used as guidance to assure the minimal procedural coverage for plant operating activities, including related maintenance activities.  Appendix A lists typical safety-related activities that should be covered by written procedures, but does not provide a complete listing of necessary procedures.  Many other activities carried out during the operation phase of a nuclear power plant require written procedures, which may or may not be applicable, because of the configuration of the nuclear power plant.  The procedures listed in Appendix A may be added to, combined, separated or deleted to conform to the applicant's procedure plan.
quirements During the Construction Phase of Nuclear Power Plants," (Green Book)                          g. Section 5.2.19. 1-The guidelines for preop erational tests, except the guideline that refers to a
                                                      1.33-2


Rev. 3 of RG 1.33, Page 4
tification in Section 5.3.9.1(2); automatic actions in run-in period for equipment. In addition to these              Section 5.3.9.1(3); immediate operator action, guidelines, the prerequisite steps for each equipment          excluding those guidelines contained in the examples,
- test should be completed prior to the commencement            in Section 5.3.9.1(4); and subsequent operator ac of the preoperational test.                                    tions in Section 5.3.9.1(5).
        h. Section 5.3.2-The guidelines that describe the content (excluding format) of procedures, except                         


==D. IMPLEMENTATION==
==D. IMPLEMENTATION==
The purpose of this section is to provide information on how applicants and licensees
for the guidelines that address (1) a separate state ment of applicability in Section 5.3.2(2), (2) inclu            The purpose of this section is to provide informa sion of references in procedures, as applicable, in          tion to applicants and licensees regarding the NRC
1 may use this guide and information regarding the NRC's plans for using this regulatory guide. In addition, it describes how the NRC staff complies with 10 CFR 50.109, "Backfitting" and any applicable finality provisions in 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."  Use by Applicants and Licensees Applicants and licensees may voluntarily
  Section 5.3.2(3), and (3) inclusion of quantitative          staff's plans for using this regulatory guide.
2 use the guidance in this document to demonstrate compliance with the underlying NRC regulations. Methods or solutions that differ from those described in this regulatory guide may be deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations.  Current licensees may continue to use Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2 for complying with the identified regulations as long as their current licensing basis remains unchanged.   Licensees may use the information in this regulatory guide for actions which do not require NRC review and approval such as changes to a facility design under 10 CFR 50.59, "Changes, Tests, and Experiments."  Licensees may use the information in this regulatory guide or applicable parts to resolve regulatory or inspection issues. Use by NRC Staff  The NRC staff does not intend or approve any imposition or backfitting of the guidance in this regulatory guide. The NRC staff does not expect any existing licensee to use or commit to using the guidance in this regulatory guide, unless the licensee makes a change to its licensing basis. The NRC staff does not expect or plan to request licensees to voluntarily adopt this regulatory guide to resolve a generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which would require the use of this regulatory guide. Examples of such unplanned NRC regulatory actions include issuance of an order requiring the use of the regulatory guide, requests for information under 10 CFR 50.54(f) as to whether a licensee intends to commit to use of this regulatory guide, generic communication, or promulgation of a rule requiring the use of this regulatory guide without further backfit consideration. During regulatory discussions on plant specific operational issues, the staff may discuss with licensees various actions consistent with staff positions in this regulatory guide, as one acceptable means of meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be considered backfitting even if prior versions of this regulatory guide are part of the licensing basis of the facility. However, unless this regulatory guide is part of the licensing basis for a facility, the staff may not represent to the licensee that the licensee's failure to comply with the positions in this regulatory guide constitutes a violatio
 
control guides in Section 5.3.2(6).                              This guide reflects current NRC practice. There fore, except in those cases in which the applicant i.. Section 5.3.9-The guideline concerning              proposes an acceptable alternative method for com emergency procedures requiring prompt implementa              plying with the specified portions of the Commis tion of immediate operator actions when required to            sion's regulations, the method described herein is prevent or mitigate the consequences of a serious              being and will continue to be used in the evaluation condition.                                                    of submittals for operating license applications until j. Section 5.3.9.1-The guidelines that describe          this guide is revised as a result of suggestions from the content (excluding format) for.' the title in Section      the public or additional staff review.
 
5.3.9.1 (1); the inclusion of symptoms to aid in iden-
                                                            1.33-3
 
APPENDIX A
                      TYPICAL PROCEDURES FOR PRESSURIZED WATER REACTORS
                                        AND BOILING WATER REACTORS
      The following are typical safety-related activities          b. Control Rod Drive System (including part that should be covered by written procedures. This            length rods)
  appendix is not intended as an inclusive listing of all c. Shutdown Cooling System needed procedures since many other activities carried d. Emergency Core Cooling System out during the operation phase of nuclear.power e. Component Cooling Water System plants should be covered by procedures not included f. Containment in this list.
 
(1) Maintaining Containment Integrity
  1. Administrative Procedures                                          (2) Special Containment Systems a. Security and Visitor Control                                      (a)  Atmosphere b. Authorities and Responsibilities for Safe Opera                    (b)  Subatmospheric tion and Shutdown                                                        (c)  Double-Wall Containment with Controlled c. Equipment Control (e.g., locking and tagging)          Interspace d. Procedure Adherence and Temporary Change                          (d)  Ice Condenser Method                                                                (3) Containment Ventilation System e. Procedure Review and Approval
                                                                      (4) Containment Cooling System f. Schedule for Surveillance Tests and Calibration g. Shift and Relief Turnover                                  g. Atmosphere Cleanup Systems h. Log Entries, Record Retention, and Review                  h. Fuel Storage Pool Purification and Cooling Sys Procedures                                                    tem i. Access to Containment                                      i. Main Steam System j. Bypass of Safety Functions and Jumper Control              j. Pressurizer Pressure and Spray Control Systems k. Maintenance of Minimum Shift Complement                    k. Feedwater System (feedwater pumps to steam and Call-In of Personnel                                      generator)
    I. Plant Fire Protection Program                              1. Auxiliary Feedwater System m. Communication System Procedures                            m. Service Water System n. Chemical and Volume Control System (includ
2. General Plant Operating Procedures                        ing Letdown/Purification System)
    a. Cold Shutdown to Hot Standby                              o. Auxiliary or Reactor Building Heating and b. Hot Standby to Minimum Load (nuclear start              Ventilation up)                                                              p. Control Room Heating and Ventilation c. Recovery from Reactor Trip                                q. Radwaste Building Heating and Ventilation d. Operation at Hot Standby                                  r. Instrument Air System e. Turbine Startup and Synchronization of                    s. Electrical System Generator                                                            (1) Offsite (access circuits)
    f. Changing Load and Load Follow (if applicable)                  (2) Onsite g. Power Operation and Process Monitoring h. Power Operation with less than Full Reactor                      (a)  Emergency Power Sources (e.g., diesel Coolant Flow                                                  generator,      batteries)
    i. Plant Shutdown to Hot Standby                                    (b)  A.C. System j. Hot Standby to Cold Shutdown                                      (c)  D.C. System k. Preparation for Refueling and Refueling                    t. Nuclear Instrument System Equipment Operation I. Refueling and Core Alterations                                (1) Source Range
                                                                    (2) Intermediate Range
3. Procedures for Startup, Operation, and Shutdown                  (3) Power Range of Safety-Related PWR Systems                                    (4) Incore System Instructions for energizing, filling, venting, drain          u. Reactor Control and Protection System ing, startup, shutdown, and changing modes of oper              v. Hydrogen Recombiner ation should be prepared, as appropriate, for the fol lowing systems:
                                                              4. Procedure for Startup, Operation, and Shutdown a. Reactor Coolant System of Safety-Related BWR Systems
                                                        1.33-4
 
Instructions for energizing, filling, venting, drain      5. Procedures for Abnormal, Offnormal, or Alarm ing, startup, shutdown, and changing modes of oper                Conditions ation should be prepared, as appropriate, for the Since these procedures are numerous and corre following systems:
                                                              spond to the number of alarm annunciators, the pro cedures are not individually listed. Each safety related annunciator should have its own written a. Nuclear Steam Supply System (Vessel and                procedure, which should normally contain (1) the Recirculating System)                                        meaning of the annunciator, (2) the source of the sig b. Control Rod Drive System                                nal, (3) the immedate action that is to occur automat c. Reactor Cleanup System                                  ically, (4) the immediate operation action, and (5) the d. Liquid Poison System (Standby Liquid Con                long-range actions.
 
trol System)
  e. Shutdown Cooling and Reactor Vessel Head Spray System
                                                              6. Procedures for Combating Emergencies and Other f. High Pressure Coolant Injection Significant Events g. Reactor Core Isolation Cooling System h. Emergency Core Cooling Systems                            a. Loss of Coolant (including significant PWR
    i. Closed Cooling Water System                            steam generator leaks) (inside and outside primary j. Containment                                            containment) (large and small, including leak-rate
      (1) Maintaining Integrity                              determination)
      (2) Containment Ventilation System                        b. Loss of Instrument Air
      (3) Inerting and deinerting                              c. Loss of Electrical Power (and/or degraded power sources)
  k. Fuel Storage Pool Purification and Cooling                d. Loss of Core Coolant Flow System                                                          e. Loss of Condenser Vacuum I. Main Steam System (reactor vessel to turbine)              f. Loss of Containment Integrity m. Turbine-Generator System                                    g. Loss of Service Water n. Condensate System (hotwell to feedwater                    h. Loss of Shutdown Cooling pumps, including demineralizers and resin                        i. Loss of Component Cooling System and Cool regeneration)                                                 ing to Individual Components o. Feedwater System (feedwater pumps to reactor              j. Loss of Feedwater or Feedwater System vessel)                                                      Failure p. Makeup System (filtration, purification, and              k. Loss of Protective System Channel water transfer)                                                  1. Mispositioned Control Rod or Rods (and rod drops)
  q. Service Water System                                      m. Inability to Drive Control Rods r. Reactor Building Heating and Ventilation                  n. Conditions Requiring Use of Emergency Bora Systems                                                      tion or Standby Liquid Control System s. Control Room Heating and Ventilation                      o. Fuel Cladding Failure or High Activity in Systems                                                      Reactor Coolant or Offgas t. Radwaste Building Heating and Ventilation                p. Fire in Control Room or Forced Evacuation of Systems                                                      Control Room u. Standby Gas Treatment System                              q. Turbine and Generator Trips v. Instrument Air System                                    r. Other Expected Transients that may be w. Electrical System                                      Applicable
        (1) Offsite (access circuits)                            s. Malfunction of Automatic Reactivity Control
        (2) Onsite                                            System t. Malfunction of Pressure Cdntrol System (a) Emergency Power Sources (e.g., diesel            u. Reactor Trip generator, batteries)                                          v. Plant Fires (b) A.C. System                                      w. Acts of Nature (e.g., tornado, flood, dam (c) D.C. System                                    failure, earthquakes)
    x. Nuclear Instrument System                                x. Irradiated Fuel Damage While Refueling y. Abnormal Releases of Radioactivity
        (1) Source Range
        (2) Intermediate Range
        (3) Power Range z. Intrusion of Demineralizer Resin Into Primary System (BWR Plants)                                      I
        (4) TIP System                                        7. Procedures for Control of Radioactivity (For limit y. Reactor Protection System                                  ing materials released to environment and limiting z. Rod Worth Minimizer                                        personnel exposure)
                                                        1.33-5


====n.      ====
a. Liquid Radioactive Waste System
1 In this section, "licensees" refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; and the term "applicants," refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts 50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52. 2  In this section, "voluntary" and "voluntarily" means that the licensee is seeking the action of its own accord, without the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.
                                                                    (1) Pressurized Water Reactors
    ( I) Collection.      Demineralizing. Filtering, Evaporating          and     Concentrating,                                (a) Containment Leak-Rate Tests and Neutralizing                                                                (b) Containment Isolation Tests
        (2) Sampling and Monitoring                                          (c) Containment Local Leak Detection Tests
        (3) Discharging to Effluents (d) Containment Heat and Radioactivity b. Solid Waste System                                        Removal Systems Tests
        (1) Spent Resins and Filter Sludge Handling                          (e) Containment Tendon Tests and
        (2) Baling Machine Operation                              Inspections
        (3) Drum Handling and Storage                                        (f) Service Water System Functional Tests (g) Main Steam Isolation Valve Tests c. PWR Gaseous Effluent System                                          (hi Fire Protection System Functional Tests
        (1) Collection. Storage. and Discharge                              (i) Boric Acid Tanks-Level Instrumenta tion Calibrations
        (2) Sampling and Monitoring
        (3) Air Ejector and Stack Monitoring                                (j) Emergency Core Cooling System Tests
        (4) Ventilation Air Monitoring (k Control Rod Operability and Scram Time Tests d. BWR Air Extraction. Offgas Treatment. and                            (1) Reactor Protection System Tests and Other Gaseous Effluent Systems                                  Calibrations
        (1) Mechanical Vacuum Pump Operation                                (m) Permissi',es-Tests and Calibrations
        (2) Air Ejector Operation                                          (n) Refueling System Circuit Tests
        (3) Packing Steam Exhauster Operation                              (o) Emergency Boration System Functional
        (4) Sampling                                            Tests
        (5) Air Ejector. Ventilation. and Stack Monitor                    (p) DNB Checks and Incore-Excore Flux Monitor Correlations e. Radiation Protection Procedures                                      (q) Emergency Power Tests (I) Access Control to Radiation Areas Includ.                      tr) Auxiliary Feedwater System Tests ing a Radiation Work Permit System                                        (s) NSSS Pressurization and Leak
        (2) Radiation Sur\e\s                                    Detection
        (3) Airborne Radioactiv.ity Monitoring                              (t) Inspection of Reactor Coolant System
        (4) Contamination Control                                Pressure Boundary
        (5) Respiratory Protection                                          (u) Inspection of Pipe Hanger Settings
        (6) Training in Radiation Protection                                (v) Control Rod Drive System Functional
        (7) Personnel Monitoring                                Tests
        (8) Bioassay Program                                              (w) Heat      Balance-Flux        Monitor
        (9) Implementation of ALARA Program                    Calibrations Ix) Pressurizer and Main Steam Safety f. Area Radiation Monitoring S)stem Operation                Valve. Tests g. Process Radiation Monitoring System                                  (y) Leak Deection Systems Tests Operation                                                                  (z) Axial and Radial Flux Pattern Determi h. Meteorological Monitoring                                  nations
8. Procedures for Control of Measuring and Test                          (aa) Area. Portable. and Airborne Radiation Equipment and for Surveillance Tests, Proce                M on it or Calibrations dures. and Calibrations                                            (bb) Process a. Procedures of a type appropriate to the cir                                            Radiation    Monitor Calibrations cumstances should be provided to ensure that tools, (cc) Environmental Monitor Calibrations gauges. instruments, controls, and other measuring                      (dd) Safety Valve Tests and testing devices are properly controlled, calib (ee) Turbine Overspeed Trip Tests rated. and adjusted at specified periods to maintain (ff) Water Storage Tanks-Level In accuracy. Specific examples of such equipment to be            strumentation Calibration calibrated and tested are readout instruments, inter                (2) Boiling Water Reactors lock permissive and prohibit circuits, alarm devices, sensors. signal conditioners, controls, protective cir                    (a) Containment Leak-Rate and Penetration cuits, and laboratory equipment.                                Leak-Rate Tests (b) Containment Isolation Tests b. Specific procedures for surveillance tests, in                      (c) Containment Vacuum Relief Valve spections. and calibrations should be written (imple            Tests menting procedures are required for each surveillance                      (d) Containment Spray System Tests test. inspection, or calibration listed in the technical (e) Standby Gas Treatment System Tests specifications):                                                (including filter tests)
                                                        1.33-6


Rev. 3 of RG 1.33, Page 5 If an existing licensee voluntarily seeks a license amendment or change and (1) the NRC staff's consideration of the request involves a regulatory issue directly relevant to this new or revised regulatory guide and (2) the specific subject matter of this regulatory guide is an essential consideration in the staff's determination of the acceptability of the licensee's request, then the staff may request that the licensee either follow the guidance in this regulatory guide or provide an eq uivalent alternative process that demonstrates compliance with the underlying NRC regulatory requirements. This is not considered backfitting as defined in 10 CFR 50.109(a)(1) or a violation of any of the issue finality provisions in 10 CFR Part 52.  Additionally, an existing applicant may be required to adhere to new rules, orders, or guidance if 10 CFR 50.109(a)(3) applies.  If a licensee believes that the NRC is either using this regulatory guide or requesting or requiring the licensee to implement the methods or processes in this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NUREG-1409, "Backfitting Guidelines" (Ref. 9) and NRC Management Directive 8.4, "Management of Facility-specific Backfitting and Information Collection" (Ref. 10).     
(f)  Service Water System Functional Tests            c. Procedures for the repair or replacement of (g) Main Steam Isolation Valve Tests              equipment should be prepared prior to beginning (h) Fire Protection System Functional Tests      work. Such procedures for major equipment that is (i)  Nitrogen Inerting System Tests                expected to be repaired or replaced during the life of (j)  Emergency Core Cooling System Tests          the plant should preferably be written early in plant (k)    Control Rod Operability and Scram Time      life. The following are examples of such procedures Tests                                                          for major equipment:
Rev. 3 of RG 1.33, Page 6 REFERENCES
            (1) Reactor Protection System Tests and                (1) Repair of PWR Steam Generator Tubes Calibrations                                                        (2) Replacement and Repair of Control Rod (m) Rod Blocks-Tests and Calibrations                Drives (n) Refueling System Circuit Tests                        (3) Replacement of Recirculation Pump Seals (o) Liquid Poison System Tests                            (4) Replacement of Important Strainers and (p) Minimum Critical Heat Flux Checks and          Filters Incore Flux Monitor Calibrations                                        Repair or Replacement of Safety Valves
3  1. U.S. Code of Federal Regulations, 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities." 
                                                                    (5)
2. U.S. Code of Federal Regulations, 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants." 
          (q) Emergency Power Tests                                (6) Repair of Incore Flux Monitoring System (r) Isolation Condenser or RCIC Tests                        Replacement of Neutron Detectors
                                                                    (7)
            (s) NSSS Pressurization and Leak Detection (t) Inspection of Reactor Coolant System            d. Procedures that could be categorized either as Pressure Boundary                                              maintenance or operating procedures should be (u) Inspection of Pipe Hanger Settings            developed for the following activities. Instructions (v) Control Rod Drive System Functional            for these activities may be included in systems Tests                                                          procedures.


3. Regulatory Guide 1.28, "Quality Assurance Program Criteria (Design and Construction), "U.S. Nuclear Regulatory Commission, Washington, DC.
(w) Heat Balance
                                                                    (1) Exercise of equipment that is normally idle (x) Autoblowdown System Tests (y) Leak Detection System Tests                  but that must operate when required (z) Axial and Radial Flux Pattern Determi                (2) Draining and Refilling Heat Exchangers
                                                                    (3) Draining and Refilling Recirculation Loop nations (aa) Area, Portable, and Airborne Radiation                (4) Draining and Refilling the Reactor Vessel Monitor Calibrations                                                (5) Draining and Refilling Steam Generators (bb) Process Radiation Monitor Calibrations                (6) Removal of Reactor Head
                                                                    (7) Disconnection and Reconnection of Wiring I        (cc) Environmental Monitor Calibrations              Penetrating Reactor Vessel Head (dd) Safety Valve Tests
                                                                    (8) Demineralizer Resin Regeneration or (ee) Turbine Overspeed Trip Test Replacement (ff) Water Storage Tanks-Level In strumentation Calibrations                                        e. General procedures for the control of mainte (gg) Reactor Building Inleakage Tests                nance, repair, replacement, and modification work should be prepared before reactor operation is begun.


4. American National Standards Institute (ANSI)/ American Nuclear Society (ANS) 3.2-2012, "Managerial, Administrative, and Quality Assurance Controls for Operational Phase of Nuclear Power Plants."
9. Procedures for Performing Maintenance                      These procedures should include information on areas such as the following:
5. American Nuclear Society
    a. Maintenance that can affect the performance
4 (ANS) 3.2/ American National Standards Institute
                                                                    (1) Method for obtaining permission and clear of safety-related equipment should be properly pre planned and performed in accordance with written              ance for operation personnel to work and for logging procedures, documented instructions, or drawings              such work and appropriate to the circumstances. Skills normally                    (2) Factors to be taken into account, including possessed by qualified maintenance personnel may              the necessity for minimizing radiation exposure to not require detailed step-by-step delineation in a pro        workmen, in preparing the detailed work procedures.
5 (ANSI) 18.7-1976, "Administrative Controls and Quality A
ssurance for the Operational Phase of Nuclear Power Plants."
6. Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC.


7. American National Standards Institute (ANSI)/American Society of Mechanical Engineers (ASME) NQA-1, "Quality Assurance Program Requirements for Nuclear Power Plants."
cedure. The following types of activities are among those that may not require detailed step-by-step writ          10. Chemical and Radiochemical Control Procedures ten procedures:                                                  Chemical and radiochemical control procedures
6  8. International Atomic Energy Agency Safety Requirement GS-R-3, "The Management Systems for Facilities and Activities," issued 2006.
      (1) Gasket Replacement                                should be written to prescribe the nature and fre
      (2) Trouble-Shooting Electrical Circuits              quency of sampling and analyses, the instructions
      (3) Changing Chart or Drive Speed Gears or            maintaining water quality within prescribed limits, Slide Wires on Recorders                                    and the limitations on concentrations of agents that may cause corrosive attack or fouling of heat-transfer b. Preventive maintenance schedules should be              surfaces or that may become sources of radiation developed to specify lubrication schedules, inspec            hazards due to activation. These procedures should tions of equipment, replacement of such items as fil          specify laboratory instructions and calibration of lab ters and strainers, and inspection or replacement of          oratory equipment. Extreme importance must be parts that have a specific lifetime such as wear rings.      placed on laboratory procedures used to determine
                                                        1.33-7


7  9. NUREG 1409, "Backfitting Guidelines," U.S. Nuclear Regulatory Commission, Washington, DC. 10. Management Directive 8.4, "Management of Facility-specific Backfitting and information Collection," U.S. Nuclear Regulatory Commission, Washington, DC.
concentration and species of radioactivity in liquids and gases prior to release, including representative sampling, validity of calibration techniques, and ade.


3 Publicly available NRC published documents are available electronically through the NRC Library on the NRC's public Web site at: http://www.nrc.gov/reading-rm/doc-collections/.  The documents can also be viewed on-line or printed for a fee in the NRC's Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, Washington, DC 20555; telephone 301-415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail pdr.resource@nrc.gov
quacy of analyses.
4 Copies of American Nuclear Society (ANS) standards may be purchased from the ANS Web site (http://www.new.ans.org/store/); or by writing to: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60526, U.S.A., Telephone 800-323-3044.


5 Copies of American National Standard s Institute (ANSI) standards may be purchased from ANSI, 1819 L Street, NW., Washington, DC 20036, on their Web site at http://webstore.ansi.org/; telephone (202) 293-8020; fax (202) 293-9287; or e-mail storemanager@ansi.org.    6 Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME, Two Park Avenue, New York, New York 10016-5990; telephone (800) 843-2763. Purchase information is available through the ASME Web-based store at http://www.asme.org/Codes/Publications/
UNITED STATES
. 7 Copies of International Atomic Energy Agency(IAEA) standards may be purchased from IAEA, Wagramerstrasse 5, P.O. Box 100, A-1400 Vienna, Austria; Telephone: +43 1 2600 22529 (or 22530). Purchase information is available through the ASME
NUCLEAR REGULATORY COMMISSION
Web-based store at http://www-pub.iaea.org/MTCD/publications/publications.asp
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Latest revision as of 11:24, 28 March 2020

Quality Assurance Program Requirements (Operation)
ML003739995
Person / Time
Issue date: 02/28/1978
From:
Office of Nuclear Regulatory Research
To:
References
RG-1.33, Rev 2
Download: ML003739995 (8)
v • d • e


Revlklon 2 U.S. NUCLEAR REGULATORY COMMISSION February 1978 REGULATORY GUIDE

OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 1.33 QUALITY ASSURANCE PROGRAM REQUIREMENTS

(OPERATION)

A. INTRODUCTION

ANS-3.2, "Administrative Controls and Quality As surance for the Operational Phase of Nuclear Power Appendix B, "Quality Assurance Criteria for Nu Plants," by the American National Standards Insti clear Power Plants and Fuel Reprocessing Plants," to tute on February 19, 1976.

10 CFR Part 50, "Licensing of Production and Utili zation Facilities," establishes quality assurance re There had been some uncertainty with regard to the quirements for the operation of nuclear power plant NRC staff's position when a regulatory guide en safety-related structures, systems, and components. dorses, as an acceptable method, the "guidelines" as This regulatory guide describes a method acceptable well as the "requirements" included in a standard.

to the NRC staff for complying with the Commis The NRC staff has evaluated the guidelines contained sion's regulations with regard to overall quality as in N18.7-1976/ANS-3.2 with respect to importance surance program requirements for the operation phase to safety. Revision I of this regulatory guide clarified of nuclear power plants. The Advisory Committee on the NRC staff's position on the "requirements" and Reactor Safeguards has been consulted concerning "guidelines" included in ANSI N18.7-1976/ANS

this guide and has concurred in the regulatory 3.2. Where conformance to the recommendations of position. this regulatory guide is indicated in an application without further qualification, this indicates the appli

B. DISCUSSION

cant will comply with the "requirements" of ANSI

N18.7-1976/ANS-3.2, as supplemented or modified Subcommittee ANS-3,' Reactor Operations, of the by the regulatory position of this guide.

American Nuclear Society Standards Committee de veloped ANSI N18.7-1972, which contained criteria for Section 1, "Scope," of ANSI NI8.7-1976/ANS

administrative controls for nuclear power plants dur 3.2 states that this standard contains criteria for ad ing operation. This standard, along with ANSI ministrative controls and quality assurance for nu N45.2-1971, "Quality Assurance Program Require clear power plants during the operational phase of ments for Nuclear Power Plants," was endorsed by plant life and that this phase is generally considered Regulatory Guide 1.33. The dual endorsement was to commence with initial fuel loading, except for cer necessary in order for the guidance contained in the tain preoperational activities. In this regard, a sepa regulatory guide to be consistent with the require rate regulatory guide addressing the quality assurance ments of Appendix B to 10 CFR Part 50; however, program for the preoperational phase will be issued.

this dual endorsement caused some confusion among Other regulatory guides may be issued or this regula users. To clarify this situation, ANSI N18.7-1972 tory guide may be revised, if necessary, to amplify was revised so that a single standard would define the the general requirements contained in this standard.

general quality assurance program "requirements"

for the operation phase. This revised standard was Appendix A to this guide has been further revised approved by the American National Standards Com as a result of additional' comments received on the mittee NI8, Nuclear Design Criteria. It was sub guide and additional staff review.

sequently approved and designated N18.7-1976/

,Copies may be obtained from American Nuclear Society, 555

-*Lines indicate substantive changes from previous issue. North Kensington Avenue, La Grange Park, Illinois 60525.

USNRC REGULATORY GUIDES Comments shoukl be sent to the Secretary of the Commission. US. Nuclear Regu latory Commission, Washington, D.C. 20555, Attention: Docketing and Service Regulatory Guides ore issued to describe and make available to the Public methods acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evaluating specific problems The guides are issued in the following ten broad divisions or postulated accidents, or to Provide guidance to applicants. Regulatory Guides awt not substitutes for regulations, and compliance with them is not required. 1. Power Reaclors 6. TProducts Methods and solutions different from those set out in the guides will be accept. 2. Research and Test Reactors 7. Trancsortation aWle if they provide a basis for the findings requisite to the issuance or continuance 3. Fuels and Materials Facilities a. Occupational Health o fa p ecr mitor license aby th eCo mni, son. 4. Environm etntl and Siting

9. Antitrust Review

5. Materials and Plant Protection t0. General Comments end suggestions for improvements in these guides we encouraged at all Requests for single copies of issued guides iwhich may be reiroduced) or for place tires, and guides will be revised, as appropriate, to accomnmodate comments and ment on an automatic distribution list for single copies of future guides in specific

"toreflect new information or experience. This guide was revised as a result of divisions should be made in writing to the US. Nuclear Regulatory Commission.

eubltantive comments received from the Public and additional staff review. Washington, D.C. 20555. Attention Director. Division of Document Coistrol.

C. REGULATORY POSITION

and will be endorsed by a regulatory guide The overall quality assurance program require upon its approval as an ANSI standard.

ments for the operation phase that are included in ANSI N18.7-1976/ANS-3.2 are acceptable to the 3. Section 4.3.4, "Subjects Requiring Independ NRC staff and provide an adequate basis for comply ent Review," Item (3) states, in part, that changes to ing with the quality assurance program requirements the technical specifications or license amendments re of Appendix B to 10 CFR Part 50, subject to the fol lated to nuclear safety are required to be reviewed by lowing: the independent review body prior to implementation.

It should be noted that proposed changes to technical I. ANSI N18.7-1976/ANS-3.2 requires the prep specifications or license amendments should be re aration of many procedures to carry out an effective viewed by the independent review body prior to their quality assurance program. Appendix A, "Typical submittal to the Commission for approval.

Procedures for Pressurized Water Reactors and Boil ing Water Reactors," to this regulatory guide should 4. Section 4.5, "Audit Program," of ANSI

be used as guidance to ensure minimum procedural N18.7-1976/ANS_3.2 states that audits of selected coverage for plant operating activities, including re aspects of operational phase activities shall be per lated maintenance activities. Appendix A lists typical formed with a frequency commensurate with their safety-related activities that should be covered by safety significance and in such a manner as to ensure written procedures but does not provide a complete that an audit of all safety-related functions is com listing of needed procedures. Many other activities pleted within a period of 2 years. In amplification of carried out during the operation phase of a nuclear this requirement, the following program elements power plant require written procedures not included should be audited at the indicated frequencies:

in Appendix A. Appendix A may also contain proce dures that are not applicable to an applicant because a. The results of actions taken to correct de of the configuration of the nuclear power plant. The ficiencies that affect nuclear safety and occur in facil procedures listed in Appendix A may be combined, ity equipment, structures, systems, or method of separated, or deleted to conform to the applicant's operation-at least once per 6 months.

procedures plan. b. The conformance of facility operation to pro

2. Throughout ANSI NI8.7-1976/ANS-3.2, other visions contained within the technical specifications documents required to be included as a part of this and applicable license conditions-at least once per standard are identified at the point of reference. The 12 months.

specific acceptability of these standards listed in ANSI N18.7-1976/ANS-3.2 has been addressed in c. The performance, training, and qualifications the latest revision of the following regulatory guides: of the facility staff-at least once per 12 months.

5. The ANSI Standard Regulatory Guide "should") guidelines (indicated by the verb of ANSI N18.7-1976/ANS-3.2 contained N45.2 1.28 in the following sections have sufficient safety impor N45.2.1 1.37 tance to be treated the same as the requirements (in N45.2.2 1.38 dicated by the verb "shall") of the standard:

N45.2.3 1.39 N45.2.4 a. Section 4.4-The guidelines concerning re

1.30

N45.2.5 view activities of the onsite operating organization,

1.94 except the guideline that refers to screening subjects N45.2.6 1.58 N45.2.8 of potential concern.

1.116 N45.2.9 1.88 b. Section 5.2.3-The guideline concerning re N45.2. 10 1.74 view and updating of standing orders.

N45.2. I I 1.64 N45.2.13 1.123 c. Section 5.2.4-The guideline concerning re N18. 1 view, updating, and cancellation of special orders.

1.8 N 18.17 1.17 d. Section 5.2.7. 1-The guidelines that address N 101.4 1.54 adequate design and testing of replacement parts.

Note: N45.2.12 is discussed in NRC e. Section 5.2.13.4-The guideline concerning documents WASH-1283, "Guidance on special handling tools and equipment.

Quality Assurance Requirements During Design and Procurement Phase of Nuclear f. Section 5.2.19(2)-The guideline for check Power Plants," (Grey Book) and WASH ing plant operating procedures during the testing pro

1309, "Guidance on Quality Assurance Re gram.

quirements During the Construction Phase of Nuclear Power Plants," (Green Book) g. Section 5.2.19. 1-The guidelines for preop erational tests, except the guideline that refers to a

1.33-2

tification in Section 5.3.9.1(2); automatic actions in run-in period for equipment. In addition to these Section 5.3.9.1(3); immediate operator action, guidelines, the prerequisite steps for each equipment excluding those guidelines contained in the examples,

- test should be completed prior to the commencement in Section 5.3.9.1(4); and subsequent operator ac of the preoperational test. tions in Section 5.3.9.1(5).

h. Section 5.3.2-The guidelines that describe the content (excluding format) of procedures, except

D. IMPLEMENTATION

for the guidelines that address (1) a separate state ment of applicability in Section 5.3.2(2), (2) inclu The purpose of this section is to provide informa sion of references in procedures, as applicable, in tion to applicants and licensees regarding the NRC

Section 5.3.2(3), and (3) inclusion of quantitative staff's plans for using this regulatory guide.

control guides in Section 5.3.2(6). This guide reflects current NRC practice. There fore, except in those cases in which the applicant i.. Section 5.3.9-The guideline concerning proposes an acceptable alternative method for com emergency procedures requiring prompt implementa plying with the specified portions of the Commis tion of immediate operator actions when required to sion's regulations, the method described herein is prevent or mitigate the consequences of a serious being and will continue to be used in the evaluation condition. of submittals for operating license applications until j. Section 5.3.9.1-The guidelines that describe this guide is revised as a result of suggestions from the content (excluding format) for.' the title in Section the public or additional staff review.

5.3.9.1 (1); the inclusion of symptoms to aid in iden-

1.33-3

APPENDIX A

TYPICAL PROCEDURES FOR PRESSURIZED WATER REACTORS

AND BOILING WATER REACTORS

The following are typical safety-related activities b. Control Rod Drive System (including part that should be covered by written procedures. This length rods)

appendix is not intended as an inclusive listing of all c. Shutdown Cooling System needed procedures since many other activities carried d. Emergency Core Cooling System out during the operation phase of nuclear.power e. Component Cooling Water System plants should be covered by procedures not included f. Containment in this list.

(1) Maintaining Containment Integrity

1. Administrative Procedures (2) Special Containment Systems a. Security and Visitor Control (a) Atmosphere b. Authorities and Responsibilities for Safe Opera (b) Subatmospheric tion and Shutdown (c) Double-Wall Containment with Controlled c. Equipment Control (e.g., locking and tagging) Interspace d. Procedure Adherence and Temporary Change (d) Ice Condenser Method (3) Containment Ventilation System e. Procedure Review and Approval

(4) Containment Cooling System f. Schedule for Surveillance Tests and Calibration g. Shift and Relief Turnover g. Atmosphere Cleanup Systems h. Log Entries, Record Retention, and Review h. Fuel Storage Pool Purification and Cooling Sys Procedures tem i. Access to Containment i. Main Steam System j. Bypass of Safety Functions and Jumper Control j. Pressurizer Pressure and Spray Control Systems k. Maintenance of Minimum Shift Complement k. Feedwater System (feedwater pumps to steam and Call-In of Personnel generator)

I. Plant Fire Protection Program 1. Auxiliary Feedwater System m. Communication System Procedures m. Service Water System n. Chemical and Volume Control System (includ

2. General Plant Operating Procedures ing Letdown/Purification System)

a. Cold Shutdown to Hot Standby o. Auxiliary or Reactor Building Heating and b. Hot Standby to Minimum Load (nuclear start Ventilation up) p. Control Room Heating and Ventilation c. Recovery from Reactor Trip q. Radwaste Building Heating and Ventilation d. Operation at Hot Standby r. Instrument Air System e. Turbine Startup and Synchronization of s. Electrical System Generator (1) Offsite (access circuits)

f. Changing Load and Load Follow (if applicable) (2) Onsite g. Power Operation and Process Monitoring h. Power Operation with less than Full Reactor (a) Emergency Power Sources (e.g., diesel Coolant Flow generator, batteries)

i. Plant Shutdown to Hot Standby (b) A.C. System j. Hot Standby to Cold Shutdown (c) D.C. System k. Preparation for Refueling and Refueling t. Nuclear Instrument System Equipment Operation I. Refueling and Core Alterations (1) Source Range

(2) Intermediate Range

3. Procedures for Startup, Operation, and Shutdown (3) Power Range of Safety-Related PWR Systems (4) Incore System Instructions for energizing, filling, venting, drain u. Reactor Control and Protection System ing, startup, shutdown, and changing modes of oper v. Hydrogen Recombiner ation should be prepared, as appropriate, for the fol lowing systems:

4. Procedure for Startup, Operation, and Shutdown a. Reactor Coolant System of Safety-Related BWR Systems

1.33-4

Instructions for energizing, filling, venting, drain 5. Procedures for Abnormal, Offnormal, or Alarm ing, startup, shutdown, and changing modes of oper Conditions ation should be prepared, as appropriate, for the Since these procedures are numerous and corre following systems:

spond to the number of alarm annunciators, the pro cedures are not individually listed. Each safety related annunciator should have its own written a. Nuclear Steam Supply System (Vessel and procedure, which should normally contain (1) the Recirculating System) meaning of the annunciator, (2) the source of the sig b. Control Rod Drive System nal, (3) the immedate action that is to occur automat c. Reactor Cleanup System ically, (4) the immediate operation action, and (5) the d. Liquid Poison System (Standby Liquid Con long-range actions.

trol System)

e. Shutdown Cooling and Reactor Vessel Head Spray System

6. Procedures for Combating Emergencies and Other f. High Pressure Coolant Injection Significant Events g. Reactor Core Isolation Cooling System h. Emergency Core Cooling Systems a. Loss of Coolant (including significant PWR

i. Closed Cooling Water System steam generator leaks) (inside and outside primary j. Containment containment) (large and small, including leak-rate

(1) Maintaining Integrity determination)

(2) Containment Ventilation System b. Loss of Instrument Air

(3) Inerting and deinerting c. Loss of Electrical Power (and/or degraded power sources)

k. Fuel Storage Pool Purification and Cooling d. Loss of Core Coolant Flow System e. Loss of Condenser Vacuum I. Main Steam System (reactor vessel to turbine) f. Loss of Containment Integrity m. Turbine-Generator System g. Loss of Service Water n. Condensate System (hotwell to feedwater h. Loss of Shutdown Cooling pumps, including demineralizers and resin i. Loss of Component Cooling System and Cool regeneration) ing to Individual Components o. Feedwater System (feedwater pumps to reactor j. Loss of Feedwater or Feedwater System vessel) Failure p. Makeup System (filtration, purification, and k. Loss of Protective System Channel water transfer) 1. Mispositioned Control Rod or Rods (and rod drops)

q. Service Water System m. Inability to Drive Control Rods r. Reactor Building Heating and Ventilation n. Conditions Requiring Use of Emergency Bora Systems tion or Standby Liquid Control System s. Control Room Heating and Ventilation o. Fuel Cladding Failure or High Activity in Systems Reactor Coolant or Offgas t. Radwaste Building Heating and Ventilation p. Fire in Control Room or Forced Evacuation of Systems Control Room u. Standby Gas Treatment System q. Turbine and Generator Trips v. Instrument Air System r. Other Expected Transients that may be w. Electrical System Applicable

(1) Offsite (access circuits) s. Malfunction of Automatic Reactivity Control

(2) Onsite System t. Malfunction of Pressure Cdntrol System (a) Emergency Power Sources (e.g., diesel u. Reactor Trip generator, batteries) v. Plant Fires (b) A.C. System w. Acts of Nature (e.g., tornado, flood, dam (c) D.C. System failure, earthquakes)

x. Nuclear Instrument System x. Irradiated Fuel Damage While Refueling y. Abnormal Releases of Radioactivity

(1) Source Range

(2) Intermediate Range

(3) Power Range z. Intrusion of Demineralizer Resin Into Primary System (BWR Plants) I

(4) TIP System 7. Procedures for Control of Radioactivity (For limit y. Reactor Protection System ing materials released to environment and limiting z. Rod Worth Minimizer personnel exposure)

1.33-5

a. Liquid Radioactive Waste System

(1) Pressurized Water Reactors

( I) Collection. Demineralizing. Filtering, Evaporating and Concentrating, (a) Containment Leak-Rate Tests and Neutralizing (b) Containment Isolation Tests

(2) Sampling and Monitoring (c) Containment Local Leak Detection Tests

(3) Discharging to Effluents (d) Containment Heat and Radioactivity b. Solid Waste System Removal Systems Tests

(1) Spent Resins and Filter Sludge Handling (e) Containment Tendon Tests and

(2) Baling Machine Operation Inspections

(3) Drum Handling and Storage (f) Service Water System Functional Tests (g) Main Steam Isolation Valve Tests c. PWR Gaseous Effluent System (hi Fire Protection System Functional Tests

(1) Collection. Storage. and Discharge (i) Boric Acid Tanks-Level Instrumenta tion Calibrations

(2) Sampling and Monitoring

(3) Air Ejector and Stack Monitoring (j) Emergency Core Cooling System Tests

(4) Ventilation Air Monitoring (k Control Rod Operability and Scram Time Tests d. BWR Air Extraction. Offgas Treatment. and (1) Reactor Protection System Tests and Other Gaseous Effluent Systems Calibrations

(1) Mechanical Vacuum Pump Operation (m) Permissi',es-Tests and Calibrations

(2) Air Ejector Operation (n) Refueling System Circuit Tests

(3) Packing Steam Exhauster Operation (o) Emergency Boration System Functional

(4) Sampling Tests

(5) Air Ejector. Ventilation. and Stack Monitor (p) DNB Checks and Incore-Excore Flux Monitor Correlations e. Radiation Protection Procedures (q) Emergency Power Tests (I) Access Control to Radiation Areas Includ. tr) Auxiliary Feedwater System Tests ing a Radiation Work Permit System (s) NSSS Pressurization and Leak

(2) Radiation Sur\e\s Detection

(3) Airborne Radioactiv.ity Monitoring (t) Inspection of Reactor Coolant System

(4) Contamination Control Pressure Boundary

(5) Respiratory Protection (u) Inspection of Pipe Hanger Settings

(6) Training in Radiation Protection (v) Control Rod Drive System Functional

(7) Personnel Monitoring Tests

(8) Bioassay Program (w) Heat Balance-Flux Monitor

(9) Implementation of ALARA Program Calibrations Ix) Pressurizer and Main Steam Safety f. Area Radiation Monitoring S)stem Operation Valve. Tests g. Process Radiation Monitoring System (y) Leak Deection Systems Tests Operation (z) Axial and Radial Flux Pattern Determi h. Meteorological Monitoring nations

8. Procedures for Control of Measuring and Test (aa) Area. Portable. and Airborne Radiation Equipment and for Surveillance Tests, Proce M on it or Calibrations dures. and Calibrations (bb) Process a. Procedures of a type appropriate to the cir Radiation Monitor Calibrations cumstances should be provided to ensure that tools, (cc) Environmental Monitor Calibrations gauges. instruments, controls, and other measuring (dd) Safety Valve Tests and testing devices are properly controlled, calib (ee) Turbine Overspeed Trip Tests rated. and adjusted at specified periods to maintain (ff) Water Storage Tanks-Level In accuracy. Specific examples of such equipment to be strumentation Calibration calibrated and tested are readout instruments, inter (2) Boiling Water Reactors lock permissive and prohibit circuits, alarm devices, sensors. signal conditioners, controls, protective cir (a) Containment Leak-Rate and Penetration cuits, and laboratory equipment. Leak-Rate Tests (b) Containment Isolation Tests b. Specific procedures for surveillance tests, in (c) Containment Vacuum Relief Valve spections. and calibrations should be written (imple Tests menting procedures are required for each surveillance (d) Containment Spray System Tests test. inspection, or calibration listed in the technical (e) Standby Gas Treatment System Tests specifications): (including filter tests)

1.33-6

(f) Service Water System Functional Tests c. Procedures for the repair or replacement of (g) Main Steam Isolation Valve Tests equipment should be prepared prior to beginning (h) Fire Protection System Functional Tests work. Such procedures for major equipment that is (i) Nitrogen Inerting System Tests expected to be repaired or replaced during the life of (j) Emergency Core Cooling System Tests the plant should preferably be written early in plant (k) Control Rod Operability and Scram Time life. The following are examples of such procedures Tests for major equipment:

(1) Reactor Protection System Tests and (1) Repair of PWR Steam Generator Tubes Calibrations (2) Replacement and Repair of Control Rod (m) Rod Blocks-Tests and Calibrations Drives (n) Refueling System Circuit Tests (3) Replacement of Recirculation Pump Seals (o) Liquid Poison System Tests (4) Replacement of Important Strainers and (p) Minimum Critical Heat Flux Checks and Filters Incore Flux Monitor Calibrations Repair or Replacement of Safety Valves

(5)

(q) Emergency Power Tests (6) Repair of Incore Flux Monitoring System (r) Isolation Condenser or RCIC Tests Replacement of Neutron Detectors

(7)

(s) NSSS Pressurization and Leak Detection (t) Inspection of Reactor Coolant System d. Procedures that could be categorized either as Pressure Boundary maintenance or operating procedures should be (u) Inspection of Pipe Hanger Settings developed for the following activities. Instructions (v) Control Rod Drive System Functional for these activities may be included in systems Tests procedures.

(w) Heat Balance

(1) Exercise of equipment that is normally idle (x) Autoblowdown System Tests (y) Leak Detection System Tests but that must operate when required (z) Axial and Radial Flux Pattern Determi (2) Draining and Refilling Heat Exchangers

(3) Draining and Refilling Recirculation Loop nations (aa) Area, Portable, and Airborne Radiation (4) Draining and Refilling the Reactor Vessel Monitor Calibrations (5) Draining and Refilling Steam Generators (bb) Process Radiation Monitor Calibrations (6) Removal of Reactor Head

(7) Disconnection and Reconnection of Wiring I (cc) Environmental Monitor Calibrations Penetrating Reactor Vessel Head (dd) Safety Valve Tests

(8) Demineralizer Resin Regeneration or (ee) Turbine Overspeed Trip Test Replacement (ff) Water Storage Tanks-Level In strumentation Calibrations e. General procedures for the control of mainte (gg) Reactor Building Inleakage Tests nance, repair, replacement, and modification work should be prepared before reactor operation is begun.

9. Procedures for Performing Maintenance These procedures should include information on areas such as the following:

a. Maintenance that can affect the performance

(1) Method for obtaining permission and clear of safety-related equipment should be properly pre planned and performed in accordance with written ance for operation personnel to work and for logging procedures, documented instructions, or drawings such work and appropriate to the circumstances. Skills normally (2) Factors to be taken into account, including possessed by qualified maintenance personnel may the necessity for minimizing radiation exposure to not require detailed step-by-step delineation in a pro workmen, in preparing the detailed work procedures.

cedure. The following types of activities are among those that may not require detailed step-by-step writ 10. Chemical and Radiochemical Control Procedures ten procedures: Chemical and radiochemical control procedures

(1) Gasket Replacement should be written to prescribe the nature and fre

(2) Trouble-Shooting Electrical Circuits quency of sampling and analyses, the instructions

(3) Changing Chart or Drive Speed Gears or maintaining water quality within prescribed limits, Slide Wires on Recorders and the limitations on concentrations of agents that may cause corrosive attack or fouling of heat-transfer b. Preventive maintenance schedules should be surfaces or that may become sources of radiation developed to specify lubrication schedules, inspec hazards due to activation. These procedures should tions of equipment, replacement of such items as fil specify laboratory instructions and calibration of lab ters and strainers, and inspection or replacement of oratory equipment. Extreme importance must be parts that have a specific lifetime such as wear rings. placed on laboratory procedures used to determine

1.33-7

concentration and species of radioactivity in liquids and gases prior to release, including representative sampling, validity of calibration techniques, and ade.

quacy of analyses.

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